Manufacturing Technology 2022, 22(6):703-708 | DOI: 10.21062/mft.2022.078

ATF Claddings after High-Temperature Steam Oxidation: WDS and Nanoindentation Studies to Characterize in-depth Material Changes

Jitka Klaisnerová ORCID...1, Leoš Křivský ORCID...1, Petra Gávelová ORCID...1, Jakub Krejčí2
1 Centrum výzkumu Řež s.r.o., Hlavní 130, 250 68 Husinec - Řež. Czech Republic
2 ÚJP PRAHA a.s., Nad Kamínkou 1345, 156 00 Zbraslav, Czech Republic

After the Fukushima accident in 2011 year, ATF nuclear fuel cladding concept was accelerated to achieve the reactor operation with the new accident tolerant structural materials. However, several designed solutions do not fulfil the accident tolerant concept but particularly increase the corrosion resistance of Zr-cladding tubes at normal operating conditions, so-called “Advaced Technology Fuel, EATF”. Cr-coated zirconium claddings following the first concept, have been the widely tested and the first full Cr-coated fuel rods have been planned to operate in LWR reactor conditions around the 2022 year. Our contribution describes the Cr-coated Zr-%1Nb cladding tube microstructure after high-temperature steam oxidation at 1200°C by means of Scanning Electron Microscopy and nanoindenta-tion methods. The article is focused on WDS line-profile studies of oxygen and chromium diffusion into the Zr-matrix. The increased Cr-diffusion with oxygen is evident causing a change in local me-chanical properties which is well-described by measurements of nanohardness and Young's modulus. In addition, the developed methodology of the WDS & nanoindentation line-analyses was also opti-mized to apply in hot-cell conditions to measure the effect of neutron-irradiation on the different coat-ings and coating/matrix interface.

Keywords: Cladding tubes, ATF, WDS, EDS, High-temperature oxidation, Nanoindentation
Grants and funding:

This article was created with the support of the Technology Agency of The Czech Republic in the Theta program as the part of the project No. TK03020169

Received: September 15, 2022; Revised: December 6, 2022; Accepted: December 7, 2022; Prepublished online: December 9, 2022; Published: January 6, 2023  Show citation

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Klaisnerová J, Křivský L, Gávelová P, Krejčí J. ATF Claddings after High-Temperature Steam Oxidation: WDS and Nanoindentation Studies to Characterize in-depth Material Changes. Manufacturing Technology. 2022;22(6):703-708. doi: 10.21062/mft.2022.078.
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    References

    1. Brachet J-Christophe, Rouesne E, Ribis J, Guilbert T, Urvoy S, Nony G, Toffolon-Masclet C, Le Saux M, Chaabane N, Palancher H, David A, Bischoff J, Augereau J, Pouillier E, "High Temperature Steam Oxida-tion of Chromium-Coated Zirconium-Based Alloys: Kinetics and Process", Corrosion Science (2020), doi: https://doi.org/10.1016/j.corsci.2020.108537 Go to original source...
    2. Gávelová, P., Halodová, P., Křivská, B., Correa, C. A., Krejčí, J., Ševeček, M., Rosnecký, V.: Microstructure of Zirconium Fuel Claddings: TEM and EBSD Studies of As-Received and Neutron-Irradiated Materias, In: Manufacturing Technology, vol. 20, no. 6, p. 720-727 Go to original source...
    3. Gávelová, P., Halodová, P., Libera, O.; Prokůpková, I., Vrtílková, V.; Krejčí, J.: Experimental verification of phase diagram calculations of Zr-based alloys after high-temperature oxidation, In: Defect and Diffusion Forum 2020, vol. 405, p. 351 - 354. Go to original source...
    4. Malerba, L.; Al Mazouzi, A.; Bertolus, M.; Cologna, M.; Efsing, P.; Jianu, A.; et al. Materials for Sustainable Nuclear Energy: A European Strategic Research and Innovation Agenda for All Reactor Generations. MDPI [Online] 2022. https://www.mdpi.com/1996-1073/15/5/1845/htm (accessed Sept 13, 2022). Go to original source...
    5. Kessedjian, G.; Jurado, B.; Aiche, M.; Barreau, G. Physics Letters B, 692th ed. [online]; pp 297-301. https://www.sciencedirect.com/science/article/pii/S0370269310009007 (accessed Sept 13, 2022). Go to original source...
    6. Krejčí J., Ševeček M., Cvrček L., Kabátová J., Manoch F.: Chromium and Chromium Nitride Coated Clad-ding for Nuclear Reactor Fuel, Conference EUROCORR 2017, Prague
    7. Yang, J.; Stegmaier, U.; Tang, Ch.; Steinbruck, M.; Grosse, M.; Wang, S.; Seifert, H. High temperature Cr-Zr interaction of two types of Cr-coated Zr alloys in inert gas environment. Journal of Nuclear Materials [Online] 2021. https://www.sciencedirect.com/science/article/pii/S0022311521000295 (accessed Sept 13, 2022). Go to original source...
    8. Abu-Shgair, K.; et al. Characterizing crystalline chromium oxide thin film growth parameters. Re-views on Advanced Materials Science [Online] 2010. https://www.researchgate.net/publication/228416200_Characterizing_crystalline_chromium_oxide_thin_film_growth_parameters (accessed July 12, 2022).
    9. In-Chul, Ch.; Brandl, Ch.; Schwaiger, R. Thermally activated dislocation plasticity in body-centered cubic chromium studied by high-temperature nanoindentation. Materials Science and Engineering [Online] 2022. https://www.sciencedirect.com/science/article/pii/S1359645417306808 (accessed Sept 12, 2022).
    10. Kewei, L.; Shuangming, L. Microstructure characterization and mechanical properties of a Laves-phase al-loy based on Cr2Nb. International Journal of Refractory Metals and Hard Materials [Online] 2013. https://www.researchgate.net/publication/271635980_Microstructure_characterization_and_mechanical_properties_of_a_Laves-phase_alloy_based_on_Cr2Nb (accessed Sept 12, 2022).
    11. Okamoto, H.; Schlesinger, M.; Mueller, E. Cr (Chromium) Binary Alloy Phase Diagrams. ASM Handbook: Volume 3, Alloy phase diagrams, 3rd ed.; 2016 Go to original source...
    12. Bublíková, P., Halodová, P., , Fokt, M., Namburi, H. K., Rosnecký, V., Procházka, J., Duchoň, J., Vojtěch, D.: Neutron irradiated reactor internals: An applied methodology for specimen preparation and Post Irradi-ation Examination by electron microscopy methods, In: Manufacturing Technology, vol. 18, no. 4, p. 545-551 Go to original source...

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